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Embrittlement of nuclear reactor pressure vessels

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Abstract

Neutron irradiation embrittlement could limit the service life of some of the reactor-pressure vessels in existing commercial nuclear-power plants. Improved understanding the of the underlying causes of embrittlement has provided regulators and power-plant operators better estimates of vessel-operating margins. This article presents an overview of embrittlement, emphasizing the status of mechanistic understanding and models, and their role in increasing the reliability of vessel-integrity assessments. Finally, a number of outstanding issues and significant opportunities, including a new fracture-toughness master-curve method, are briefly described.

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References

  1. Rules and Regulations Title 10 Code of Federal Regulations Part 50 Fracture Toughness Requirement (Washington, D.C.: U.S. Government Printing Office, U.S. Nuclear Regulatory Commission, 1986).

  2. Boiler and Pressure Vessel Code Section III App. G Protection Against Nonductile Fracture (New York: American Society of Mechanical Engineers, 1986).

  3. Regulatory Guide 1.99-Rev 2: Radiation Embrittlement to Reactor Pressure Vessel Materials (Washington, D.C.: U.S. Government Printing Office, U.S. Nuclear Regulatory Commission, 1988).

  4. E.D. Eason, J.E. Wright, and G.R. Odette, Improved Embrittlement Correlations for Reactor Pressure Vessel Steels, NUREG/CR-6551 (Washington, D.C.: U.S. Government Printing Office, U.S. Nuclear Regulatory Commission, 1998).

    Google Scholar 

  5. Standard Test Method for Determination of Reference Temperature To for Ferritic Steels in the Transition Range, ASTM E1921-97 Annual Book of ASTM Standards 03.01 (West Conshohocken, PA: ASTM, 1998), pp. 1068–1084.

  6. G.R. Odette and G.E. Lucas, “Recent Progress in Understanding Reactor Pressure Vessel Embrittlement,” Rad. Effects and Defects in Solids, 144 (1998), pp. 189–231.

    CAS  Google Scholar 

  7. E.D. Eason et al., “Embrittlement Recovery Due to Annealing of Reactor Pressure VesselSteels,” NuclEngDes, 179(3)(1998), pp. 257–265.

    CAS  Google Scholar 

  8. G.R. Odette et al., “Multiscale-Multiphysics Modeling of Radiation Damaged Materials: Embrittlement of Pressure Vessel Steels,” MRS Bulletin, 26 (3) (2001), pp. 176–181.

    CAS  Google Scholar 

  9. G.R. Odette and G.E. Lucas, “Irradiation Embrittlement of Reactor Pressure Vessel Steels: Mechanisms, Models and Data Correlations,” Radiation Embrittlement of Reactor Pressure Vessel Steels—An International Review, ASTM STP 909, ed. L.E. Steele (Philadelphia, PA: ASTM, 1986), pp. 206–241.

    Google Scholar 

  10. G.R. Odette, P.M. Lombrozo, R.A. Wullaert, “The Relationship Between Irradiation Hardening and Embrittlement of Pressure Vessel Steels,” Effects of Radiation on Materials—12th Int. Symp., ASTM STP 870, ed. F.A. Garner and J.S. Perrin (Philadelphia, PA: ASTM, 1985), pp. 840–860.

    Google Scholar 

  11. R.E. Stoller, G.R. Odette, and B.D. Wirth, “Primary Damage Formation in BCC Iron,” J. Nucl. Mater., 251 (1997), pp. 49–60.

    Article  CAS  Google Scholar 

  12. B.D. Wirth and G.R. Odette, “Kinetic Lattice Monte Carlo Simulations of Cascade Aging in Dilute Iron-Copper Alloys,” Microstructural Processes in Irradiated Materials, MRS Symp. Proc. 540, ed. J Zinkle et al. (Warrendale, PA: MRS, 1999), pp. 637–642.

    Google Scholar 

  13. B.D. Wirth et al., “Dislocation Loop Structure Energy and Mobility of Self-interstitial Atom Clusters in BCC Iron,” J. Nucl. Mater., 276 (2000), pp. 33–40.

    Article  CAS  Google Scholar 

  14. G.F. Solt et al., “Irradiation Induced Precipitation in Model Alloys with Systematic Variations of Cu, N and P Content: A Small Angle Neutron Scattering Study,” Effects of Radiation on Materials—16th Int. Symp., ASTM STP 1175, ed. A.S Kumar et al. (West Conshohocken, PA: ASTM, 1993), pp. 444–462.

    Google Scholar 

  15. W.J. Phythian and C.A. English, “Microstructural Evolution in Reactor Pressure-Vessel Steels,” J. Nucl. Mater., 205 (1993), pp. 162–177.

    Article  CAS  Google Scholar 

  16. T.J. Williams and W.J. Phythian, “Electron Microscopy and Small Angle Neutron Scattering Study of Precipitates in Low-Alloy Submerged-Arc Welds,” Effects of Radiation on Materials—17th Int. Symp., ASTM 1270, ed. D.S. Gelles et al. (West Conshohocken, PA: ASTM, (1996), pp. 191–205.

    Google Scholar 

  17. M.L. Jenkins, “Characterization of Radiation Damage Microstructures by TEM,” J. Nucl. Mater., 216 (1994), pp. 125–156.

    Article  Google Scholar 

  18. M.K. Miller, P. Pareige, and M.G. Burke, “Understanding Pressure Vessel Steels: An Atom Probe Perspective,” Materials Characterization, 44 (2000), pp. 235–254.

    Article  CAS  Google Scholar 

  19. B.D. Wirth et al., submitted to Phys. Rev. B.

  20. G.R. Odette, “Radiation Induced Microstructural Evolution in Reactor Pressure Vessel Steels,” Microstructural Evolution During Irradiation, MRS. Symp. Proc. 373, ed. I. Robertson et al. (Pittsburgh, PA: MRS, 1995), pp. 137–148.

    Google Scholar 

  21. G.R. Odette and B.D. Wirth, “A Computational Microscopy Study of Nanostructural Evolution in Irradiated Pressure Vessel Steels,” J. Nucl. Mater., 251 (1997), pp. 157–171.

    Article  CAS  Google Scholar 

  22. D.L. Liu et al., “A Lattice Monte Carlo Simulation of Nanophase Compositions and Structures in Irradiated Pressure Vessel Fe-Cu-Ni-Mn-Si Steels,” Mater. Sci. Eng. A—Struct., 238 (1) (1997), pp. 202–209.

    Article  Google Scholar 

  23. G.R. Odette, G.E. Lucas, and D. Klingensmith “On the Effects of Neutron Flux and Composition on Hardening of Reactor Pressure Vessel Steels and Model Alloys,” Microstructural Evolution During Irradiation, MRS. Symp. Proc., ed. G.E. Lucas et al. (Pittsburgh, PA: MRS, 2001), in press.

    Google Scholar 

  24. Q. Yu et al., “Hardening and Microstructure of Model Reactor Pressure Vessel Steel Alloys Using Proton Irradiation,” Microstructural Evolution During Irradiation, MRS. Symp. Proc., ed. G.E. Lucas et al. (Pittsburgh, PA: MRS, 2001), in press.

    Google Scholar 

  25. S.B. Fisher and J.T. Buswell, “A Model For PWR Pressure-Vessel Embrittlement,” Int. J. Pressure Vessels and Piping, 27 (2) (1987), pp. 91–135.

    Article  Google Scholar 

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Authors and Affiliations

  1. the Department of Mechanical and Environmental Engineering, the University of California, Santa Barbara

    G. R. Odette (professor) & G. E. Lucas (professor)

Authors
  1. G. R. Odette
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  2. G. E. Lucas
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Additional information

For more information, contact G.R. Odette, University of California at Santa Barbara, Department of Mechanical and Environmental Engineering, Santa Barbara, CA 93106; (805) 893-3525; fax (805) 893-8651; e-mail odette@engineering.ucsb.edu.

Editor’s Note: A hypertext-enhanced version of this article can be found at www.tms.org/pubs/journals/JOM/0107/Odette-0107.html

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Odette, G.R., Lucas, G.E. Embrittlement of nuclear reactor pressure vessels. JOM 53, 18–22 (2001). https://doi.org/10.1007/s11837-001-0081-0

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  • DOI: https://doi.org/10.1007/s11837-001-0081-0

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